ML14064A329
| ML14064A329 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 02/26/2014 |
| From: | Westinghouse |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14064A318 | List: |
| References | |
| ET 14-0008 LTR-LIS-13-557, Rev 0 | |
| Download: ML14064A329 (43) | |
Text
Enclosure I1to ET 14-0008 Westinghouse Electric Company LLC, LTR-LIS-13-557 Revision 0, NP-Attachment, "Application of Westinghouse Best-Estimate Large Break LOCA Methodology to the Wolf Creek Generating Station" (Non-Proprietary)
(34 pages)
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 1 of 34 NP-Attachment to LTR-LIS-13-557, Rev 0 Application of Westinghouse Best-Estimate Large Break LOCA Methodology to the Wolf Creek Generating Station
@ 2014 Westinghouse Electric Company LLC All Rights Reserved
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 2 of 34 APPLICATION OF WESTINGHOUSE BEST-ESTIMATE LARGE BREAK LOCA METHODOLOGY TO THE WOLF CREEK GENERATING STATION 1.0 Introduction Westinghouse obtained generic Nuclear Regulatory Commission (NRC) approval of its original topical report describing best-estimate (BE) large break loss-of-coolant accident (LBLOCA) methodology in 1996. NRC approval of the methodology is documented in the NRC safety evaluation report appended to the topical report [I].
Westinghouse subsequently underwent a program to revise the statistical approach used to develop the peak cladding temperature (PCT) and oxidation results at the 95th percentile. This method is still based on the Code Qualification Document (CQD) methodology [1] and follows the steps in the Code Scaling Applicability and Uncertainty (CSAU) methodology. However, the uncertainty analysis (Element 3 in CSAU) is replaced by a technique based on order statistics.
The Automated Statistical Treatment of Uncertainty Method (ASTRUM) methodology replaces the response surface technique with a statistical sampling method where the uncertainty parameters are simultaneously sampled for each case.
The approved ASTRUM evaluation model is documented in WCAP-16009-P-A [2].
A plant-specific adaptation of the Westinghouse ASTRUM BE LBLOCA Evaluation Model was performed for the large break LOCA accident analysis for Wolf Creek Generating Station (WCGS), and is summarized herein. Further discussion of the plant-specific adaptation is provided in Sections 2.0 and 3.0 herein. Both Wolf Creek Nuclear Operating Corporation (WCNOC) and its analysis vendor (Westinghouse) have interface processes which identify plant configuration changes potentially impacting safety analyses. These interface processes, along with vendor internal processes for assessing evaluation model changes and errors, are used to identify the need for LOCA analysis impact assessments.
2.0 Method of Thermal Analysis for WCGS When the Final Acceptance Criteria (FAC) governing the loss-of-coolant accident (LOCA) for Light Water Reactors was issued in 10 CFR 50.46 [3], both the NRC and the industry recognized that stipulations of Appendix K were highly conservative. That is, using the then accepted analysis methods, the performance of the Emergency Core Cooling System (ECCS) would be conservatively underestimated, resulting in predicted PCTs much higher than expected. At that time, however, the degree of conservatism in the analysis could not be quantified. As a result, the NRC began a large-scale confirmatory research program with the following objectives:
- Identify, through separate effects and integral effects experiments, the degree of conservatism in those models required in the Appendix K rule. In this fashion, those areas in which a purposely prescriptive approach was used in the Appendix K rule could be quantified with additional data so that a less prescriptive future approach might be allowed.
- Develop improved thermal-hydraulic computer codes and models so that more accurate and realistic accident analysis calculations could be performed. The purpose of this research was to develop an accurate predictive capability so that the uncertainties in the ECCS performance and the degree of conservatism with respect to the Appendix K limits could be quantified.
Since that time, the NRC and the nuclear industry have sponsored reactor safety research programs directed at meeting the above two objectives. The overall results have quantified the conservatism in the Appendix K rule for LBLOCA analyses and confirmed that some relaxation of the rule can be made without a loss in safety to the public. It was also found that some plants were being restricted in operating flexibility by overly conservative Appendix K requirements. In recognition of the Appendix K conservatism that was being quantified by the research programs, the NRC adopted an interim approach for evaluation methods. This interim approach is described in SECY-83-472 [4]. The SECY-83-472 approach retained those features of Appendix K that were legal requirements, but permitted applicants to use best-estimate thermal-hydraulic models in their ECCS evaluation model. Thus, SECY-83-472 represented an important step in basing licensing decisions on realistic calculations, as opposed to those calculations prescribed by Appendix K.
In 1988, the NRC Staff amended the requirements of 10 CFR 50.46 and Appendix K, "ECCS Evaluation Models," to permit the use of a realistic evaluation model to analyze the performance of the ECCS during a hypothetical LOCA. This decision was based on an improved understanding of LOCA thermal-hydraulic phenomena gained by extensive research programs.
Under the amended rules, best-estimate thermal-hydraulic models may be used in place of models with Appendix K features.
The rule change also requires, as part of the LOCA analysis, an assessment of the uncertainty of the best-estimate calculations. It further requires that this analysis uncertainty be included when comparing the results of the calculations to
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 3 of 34 the prescribed acceptance criteria of 10 CFR 50.46. Further guidance for the use of best-estimate codes is provided in Regulatory Guide 1.157 [5].
To demonstrate use of the revised ECCS rule, the NRC and its consultants developed a method called the CSAU evaluation methodology [6]. This method outlined an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifying the uncertainties in a LOCA analysis.
A LBLOCA evaluation methodology for three and four loop pressurized water reactor (PWR) plants based on the revised 10 CFR 50.46 rules was developed by Westinghouse with the support of EPRI and Consolidated Edison and has been approved by the NRC [1].
More recently, Westinghouse developed an alternative uncertainty methodology called ASTRUM, which stands for Automated Statistical Treatment of Uncertainty Method [2]. This method is still based on the CQD methodology and follows the steps in the CSAU methodology. However, the uncertainty analysis (Element 3 in the CSAU) is replaced by a technique based on order statistics. The ASTRUM methodology replaces the response surface technique with a statistical sampling method where the uncertainty parameters are simultaneously sampled for each case. The ASTRUM methodology has received NRC approval for referencing in licensing applications in WCAP-16009-P-A [2]. The ASTRUM methodology requires the execution of 124 transients to determine a bounding estimate of the 95th percentile (with 95% confidence level) of the PCT, Maximum Local Oxidation (MLO), and Core Wide Oxidation (CWO) to satisfy the 10 CFR 50.46 criteria with regard to PCT, MLO, and CWO.
This analysis is in accordance with the applicability limits and usage conditions defined in Section 13-3 of WCAP-16009-P-A [2] as applicable to the ASTRUM methodology. Section 13-3 of WCAP-16009-P-A [2] was found to acceptably disposition each of the identified conditions and limitations related to WCOBRA/TRAC and the CQD uncertainty approach per Section 4.0 of the ASTRUM Final Safety Evaluation Report (SER) appended to this WCAP.
The WCGS BE LBLOCA analysis includes consideration of the effects of fuel thermal conductivity degradation (TCD),
consistent with NRC expectations as noted in Information Notice 2011-21 [8]. The NRC-approved ASTRUM methodology
[2] is based on the PAD 4.0 fuel performance code [7], which does not explicitly consider fuel TCD with burnup.
Additionally, the NRC-approved ASTRUM methodology [2] only considers fuel in its first cycle of irradiation. The WCGS BE LBLOCA analysis considers both fuel performance data explicit modeling TCD and consideration of higher burnup fuel.
A detailed explanation of the TCD modeling and WCGS analysis approach is provided in Section 3.0 herein.
The WCGS BE LBLOCA analysis was performed using heat transfer multipliers different than those used in the as-approved ASTRUM methodology [2]. These heat transfer multiplier distribution changes have been reported to the NRC in [9].
3.0 Inclusion of the Effects of TCD on the BE LBLOCA Analysis for WCGS A description of the approach used for the modeling of TCD is provided in the following subsections.
3.1 Input Parameters, Plant Model, and Assumptions A plant-specific WCOBRA/TRAC model was developed for use in the WCGS BE LBLOCA analysis. The plant nodalization figures are provided as Figures 19 through 23. Figure 19 provides a vertical view of the vessel noding diagram, Figures 20 through 22 provide horizontal views of each section of the vessel noding diagram, and Figure 23 provides the loop noding diagram.
The results of the CCTF Test 62, UPTF Test 6, and UPTF Test 25A simulations for the approved CQD methodology are described in Sections 14-2-6-1, 14-4-5 through 14-4-9, and 14-4-11 of the CQD [1], respectively. These simulations were then extended for the ASTRUM EM in Appendix B of [2]. The as-approved methodology includes modeling a single downcomer channel stack per loop.
At times, recent Westinghouse BE LBLOCA analyses have modeled three downcomer channel stacks per loop.
Ia~c
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 4 of 34 Table I lists the major plant parameter assumptions used in the BE LBLOCA analysis for WCGS.
The peaking factor values utilized in the WCGS BE LBLOCA analysis are shown in Table 2. In order to mitigate the impact of the increasing effect of fuel TCD with bumup, burndown credit for the hot rod and hot assembly is taken for higher bumup fuel in the second and third cycle of operation. The usage of the Table 2 peaking factors for modeling higher bumup fuel is further discussed in Section 3.3.
3.2 Determination of Plant Operating Conditions Prior to performing the uncertainty analysis, confirmatory calculations are performed to determine the limiting settings for a subset of parameters. The evaluated parameters are discussed in Section 11-3-1of the ASTRUM topical [2]. The results of these calculations are used to define the Reference Transient, which is used as a starting point for the uncertainty analysis and in the calculation of a conservatively low containment backpressure. The Reference Transient and the conservatively low containment backpressure study did not consider certain changes being included in the uncertainty study phase of the WCGS BE LBLOCA analysis described herein, and so were re-evaluated to assure continued applicability. The following items were considered when evaluating the Reference Transient applicability:
Consideration of TCD Iax Consideration of Input Changes The WCGS BE LBLOCA analysis described herein incorporated new safety injection (SI) flow data which was not considered in the original confirmatory calculations. Since the SI injection time is after the limiting PCT time for the Reference Transient run, the change in the flow rate would not impact the PCT results for the run. Additionally, based on the change in the flows and the run behavior for the non-limiting confirmatory calculations, it is unlikely that a different configuration would become more limiting than the original Reference Transient.
]a*c Consideration of New Code Versions The WCGS BE LBLOCA analysis described herein uses newer code versions than were used for the confirmatory calculations. As the code version changes would impact all runs equally, the use of different code versions would not invalidate the choice of the Reference Transient.
Containment Backpressure Applicability The conservatively low containment backpressure remains bounding since the items noted above would either have a negligible effect on the calculated containment backpressure, or, in the case of modeling fuel TCD, would increase core stored energy, thus increasing the containment pressure.
Conclusions Based on the arguments presented in this section, it is confirmed that the Reference Transient and the conservatively low containment backpressure study are appropriate for the uncertainty study phase of the WCGS BE LBLOCA analysis described herein.
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 5 of 34 3.3 Description of BE LBLOCA Analysis and Evaluations The NRC-approved Westinghouse BE LBLOCA ASTRUM methodology [2] is based on the PAD 4.0 fuel performance code
[7]. PAD 4.0 was licensed without explicitly considering fuel TCD with burnup. Explicit modeling of TCD in the fuel performance code leads directly to increased fuel temperatures (pellet radial average temperature) as well as other fuel performance related effects beyond BOL. Since PAD provides input to the LBLOCA analysis, this will tend to increase the stored energy at the beginning of the simulated LBLOCA event. This in turn leads to an increase in PCT if there is no provision to credit off-setting effects.
The WCGS BE LBLOCA analysis also credited peaking factor bumdown to evaluate higher burnup fuel in its second/third cycle of irradiation. Evaluation of fuel in its second/third cycle of irradiation is beyond the first cycle considered in the approved ASTRUM Evaluation Model [2], but was considered in the WCGS BE LBLOCA analysis when explicitly modeling TCD to demonstrate that conformance to the acceptance criteria is met for the second/third cycle fuel. Physically, accounting for TCD leads to an increase in fuel temperature as the fuel is burned, while accounting for peaking factor burndown leads to a reduction in fuel temperature as the fuel is burned. The compensating nature of these phenomena is considered in the WCGS BE LBLOCA analysis in order to appropriately capture the effect of TCD.
The WCGS BE LBLOCA analysis was performed by running all 124 cases in both the first cycle and the second cycle.
Therefore the same non-parametric order statistics singular statement of a 95th percentile at the 95-percent confidence joint probability for PCT, MLO and CWO of an ASTRUM BELOCA analysis is ensured for the WCGS BE LBLOCA analysis.
[
IFBA Fuel Iac Treatment of Burnup and Fuel in its Second and Third Cycle of Irradiation The NRC-approved Westinghouse ASTRUM Methodology [2] assumes a LOCA to be
]*C The burnup sampling method is discussed in detail in Section 11-2-2 of the ASTRUM Topical [2]. Although a small amount of peaking factor burndown credit might have been available for some of the higher burnup cases of the first cycle runset (see Table 2),
no credit was taken.
Ia~c For each of the second cycle runs, [
Ia~c
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 6 of 34
[ ]'c Since the analyzed burnup range covers up to the [ ]* design limit, fuel in its first, second or third cycles of operation is addressed.
Hot Rod Burnup (GWD/MTU) ax Overall Limiting PCT Case First Cycle Study Second Cycle Study 4.0 Description of a Large Break LOCA Transient Before the break occurs, the reactor coolant system (RCS) is assumed to be operating normally at full power in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. A large break is assumed to open instantaneously in one of the main RCS cold leg pipes.
Immediately following the cold leg break, a rapid system depressurization occurs along with a core flow reversal due to a high discharge of sub-cooled fluid into the broken cold leg and out of the break. The fuel rods go through departure from nucleate boiling (DNB) and the cladding rapidly heats up, while the core power decreases due to voiding in the core. The hot water in the core, upper plenum, and upper head flashes to steam, and subsequently, the cooler water in the lower plenum and downcomer begins to flash. Once the system has depressurized to the accumulator pressure, the accumulator begins to inject cold borated water into the intact cold legs. During the blowdown period, a portion of the injected ECCS water is calculated to be bypassed around the downcomer and out of the break. The bypass period ends as the system pressure continues to decrease and approaches the containment pressure, resulting in reduced break flow and consequently, reduced core flow.
As the refill period begins, the core continues to heat up as the vessel begins to fill with ECCS water. This phase continues until the lower plenum is filled, the bottom of the core begins to reflood, and entrainment begins.
During the reflood period, the core flow is oscillatory as ECCS water periodically rewets and quenches the hot fuel cladding, which generates steam and causes system repressurization. The steam and entrained water must pass through the vessel upper plenum, the hot legs, the steam generators, and the reactor coolant pumps before it is vented out of the break. This flow path resistance is overcome by the downcomer water elevation head, which provides the gravity driven reflood force.
The pumped cold leg injection ECCS water aids in the filling of the vessel and downcomer, which subsequently supplies water to maintain the core and downcomer water levels and complete the reflood period.
5.0 WCGS BE LBLOCA Analysis Results The major plant parameter assumptions used in the WCGS BE LBLOCA analysis are provided in Table 1. The results of the WCGS BE LBLOCA analysis are summarized in Table 3. Table 4 contains a sequence of events for the limiting PCT transient.
The scatter plot presented in Figure 1 shows the effect of the effective break area on the analysis PCT. The effective break area is calculated by multiplying the discharge coefficient (CD) with the sampled value of the break area, normalized to the cold leg cross sectional area. Figure 1 is provided to show that the break area is a significant contributor to the variation in PCT.
Figure 2 shows the predicted HOTSPOT and WCOBRA/TRAC PCT for the limiting PCT case. The HOTSPOT PCT plot includes local uncertainties applied to the Hot Rod. Figure 3 provides the PCT elevation on the Hot Rod for the PCT limiting case. Figure 4 provides the WCOBRA/TRAC PCT for all the fuel rods in the core channels.
Figures 5 through 16 illustrate the key major response parameters for the limiting PCT transient. The containment backpressure utilized for the WCGS BE LBLOCA analysis compared to the calculated containment backpressure is provided in Figure 17. The worst single failure for the WCGS BE LBLOCA analysis is the loss of one train of ECCS Safety Injection (consistent with the ASTRUM Topical [2]); however, all containment systems which would reduce containment pressure are modeled for the WCGS BE LBLOCA containment backpressure calculation.
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 7 of 34 6.0 10 CFR 50.46 Requirements It must be demonstrated that there is a high level of probability that the following limits set forth in 10 CFR 50.46 are met:
(b)(1) The limiting PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level. Since the resulting PCT for the limiting case is 19001F for WCGS, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(1), i.e., "Peak Cladding Temperature less than 2200'F," is demonstrated. The result is shown in Table 3.
(b)(2) The maximum cladding oxidation corresponds to a bounding estimate of the 95th percentile MLO at the 95-percent confidence level. Since the resulting MLO for the limiting case is 4.21 percent for WCGS, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(2), i.e., "Maximum Local Oxidation of the cladding less than 17 percent of the total cladding thickness before oxidation," is demonstrated. The result is shown in Table 3.
(b)(3) The limiting core-wide oxidation corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. While the limiting MLO is determined based on the single Hot Rod, the CWO value can be conservatively chosen as that calculated for the limiting Hot Assembly Rod (HAR) when there is significant margin to the regulatory limit. The limiting HAR total maximum oxidation is 0.1352 percent for WCGS. Thus, a detailed CWO calculation is not needed because the calculations would include many lower power assemblies and the outcome would always be less than the limiting HAR total maximum oxidation.
Therefore, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(3), i.e., "Core-Wide Oxidation less than 1 percent of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume," is demonstrated. The result is shown in Table 3.
(b)(4) 10 CFR 50.46 acceptance criterion (b)(4) requires that the calculated changes in core geometry are such that the core remains amenable to cooling. This criterion has historically been satisfied by adherence to criteria (b)(1) and (b)(2), and by assuring that fuel deformation due to combined LOCA and seismic loads is specifically addressed. It has been demonstrated that the PCT and MLO limits remain in effect for BE LBLOCA applications. The grid crush calculations currently in place for WCGS remain unchanged with the application of the ASTRUM methodology [2], and this conclusion is unaffected by the modeling of fuel TCD, therefore, acceptance criterion (b)(4) is satisfied.
(b)(5) 10 CFR 50.46 acceptance criterion (b)(5) requires that long-term core cooling be provided following the successful initial operation of the ECCS. Long-term cooling (LTC) is dependent on the demonstration of continued delivery of cooling water to the core. The actions, automatic or manual, that are currently in place at WCGS to maintain long-term cooling remain unchanged with the application of the ASTRUM methodology
[2].
When considering the explicit modeling of fuel TCD, the primary impact that is potentially important to LTC is an increase in initial fuel pellet temperature. This in turn leads to a higher amount of stored energy at the initiation of the LOCA event. Initial stored energy is not important to LTC evaluations as these evaluations only consider decay heat removal during the sump recirculation phase of ECCS operation. The increased stored energy in the fuel due to higher fuel pellet temperature is a short term effect that does not persist into the LTC phase of ECCS performance evaluations; therefore, the heat source remains limited to decay heat for LTC evaluations. Consequential impacts of higher fuel pellet temperature such as higher fuel rod internal pressure also have no impact on LTC evaluations as fuel cladding temperatures are maintained well below the threshold for cladding rupture such that cladding burst and blockage does not occur during LTC. As such, it is shown that no additional LTC analysis is required to assess TCD for WCGS, and therefore, acceptance criterion (b)(5) is satisfied.
Based on the ASTRUM analysis results (see Table 3), it is concluded that WCGS continues to maintain compliance to the limits prescribed by 10 CFR 50.46.
7.0 References
[1] Bajorek, S. M., et. al., "Code Qualification Document for Best Estimate LOCA Analysis," WCAP-12945-P-A, Volume 1, Revision 2 and Volumes 2 through 5, Revision 1, and WCAP-14747 (Non-Proprietary), March 1998.
Westinghouse Non-Proprietary Class 3 LTR-LIS-I 3-557 Revision 0, NP-Attachment Page 8 of 34
[2] Nissley, M. E., et. al., "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," WCAP- 16009-P-A and WCAP- 16009-NP-A (Non-Proprietary),
January 2005.
[3] "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume 39, Number 3, January 1974.
[4] Information Report from W.J. Dircks to the Commissioners, "Emergency Core Cooling System Analysis Methods," SECY-83-472, November 1983.
[5] "Best Estimate Calculations of Emergency Core Cooling System Performance," Regulatory Guide 1.157, USNRC, May 1989.
[6] Boyack, B., et. al., "Quantifying Reactor Safety Margins: Application of Code Scaling Applicability and Uncertainty (CSAU) Evaluation Methodology to a Large Break Loss-of-Coolant-Accident," NUREG/CR-5249, 1989.
[7] Westinghouse Report WCAP- 15063-P-A, Revision I with Errata, "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," July 2000. (Westinghouse Proprietary Class 2)
[8] NRC Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation," December 2011. (NRC ADAMS Accession Number ML113430785).
[9] LTR-NRC-14-1, "Westinghouse Report for Changes to the Heat Transfer Multiplier Uncertainty Distributions Used in the Approved BE LBLOCA Methodology," January 2014.
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 9 of 34 Table 1 Major Plant Parameter Assumptions Used in the WCGS BE LBLOCA Analysis Parameter Value PlantPhysical Description
- SG Tube Plugging <10%
PlantInitial Operating Conditions
" Reactor Power 1<3565 MWt (+ 2% uncertainty)
- Peaking Factors see Table 2
" Axial Power Distribution see Figure 18 FluidConditions
- TAVG 570.7 - 4.0°F < TAVG < 588.4 + 4.0°F
" Pressurizer Pressure 2250 - 50 psia < PRcs < 2250 + 50 psia
- Reactor Coolant Flow > 90,324 gpm/loop
" Accumulator Temperature 50OF < TACC < 120OF
" Accumulator Pressure 582.8 psia < PACC < 696.6 psia
" Accumulator Water Volume 818 ft3 < VACC < 881 ft3
" Accumulator Boron Concentration > 2300 ppm Accident Boundary Conditions
" Single Failure Assumptions Loss of one ECCS train
" Safety Injection Flow Minimum
- Safety Injection Temperature 37 0 F _<TSI _ 120OF
- Safety Injection Initiation Delay Time 27 sec (with offsite power) 39 sec (without offsite power)
- Containment Pressure Bounded (minimum); see Figure 17
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 10 of 34 Table 2 WCGS BE LBLOCA Summary of Peaking Factor Burndown Utilized Hot Rod Burnup FdH FQ Transient FQ Steady-state (GWD/MTU) (with uncertainties) (with uncertainties) (without uncertainties) 0 1.65 2.50 2.121 30 1.65 2.50 2.121 60 1.525 2.25 1.906 62 1.525 2.25 1.906 Note: Full Table is used for second cycle, values at Bumup=0 were utilized for first cycle Table 3 WCGS BE LBLOCA Analysis Results 10 CFR 50.46 Requirement Value Criteria 95/95 Peak Cladding Temperature ('F) 1900 < 2200 95/95 Maximum Local Oxidation (%) 4.21 < 17 95/95 Core-Wide Oxidation (%) 0.1352 <1
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 11 of 34 Table 4 WCGS BE LBLOCA Analysis Limiting PCT Case Sequence of Events Event Time After Break (see)
Start of Transient 0.0 Safety Injection Signal 6.0 Accumulator Injection Begins 10.0 End of Blowdown 21.0 Bottom of Core Recovery 30.0 Accumulator Begins to Empty 34.0 Safety Injection Begins 45.0 PCT Occurs 70.0 Quench Time (WCOBRA/TRAC) 310.0 End of Transient 350.0
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 12 of 34 PCT vs. (CD A) (A 124 Cases)
A M PCT A PCiT DEG D
SPL 0 0 PCT DEGCL [deg F]
2000 1*
0 1800- m U U 1600- U U
U n 1400-AA hrmm AA&A~A A~
1200- M&*1 A 1000-i i i t i i i i i JiA Qnn . . . . .
uuu 0 0.5 1 1.5 2 2.5 CD
(CD = Discharge Coefficient, Abreak = Break Area, ACL = Cold Leg Area)
I
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 13 of 34 Wolf Creek ASTRUM TCD BELOCA Ana ysis HOTSPOT PCT AT THE LIMITING ELEVATION
. WC/T PCT i-E 0 50 100 150 200 250 300 350 Time After Break (s) 276688302 Figure 2 WCGS Limiting PCT Case - HOTSPOT Clad Temperature at the Limiting Elevation and WCOBRA/TRAC PCT
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 14 of 34 Wolf Creek PCT ASTRUM TCD BELOCA Ana ysis LOCATION 11 10-9-
CO G_>
1.
7-0 50 100 150 200 250 300 350 Time After Break (S) 276688302 Figure 3 WCGS Limiting PCT Case - WCOBRAITRAC PCT Location
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 15 of 34 Wo f Creek ASTRUM TCD BELOCA Ana ysis HOT ROD PEAK CLADDING TEMPERATURE
.. - - . HOT ASSEMBLY PEAK CLADDING TEMPERATURE GUIDE TUBES PEAK CLADDING TEMPERATURE SUPPORT COLUMNS PEAK CLADDING TEMPERATURE
.. . . LOW-POWER REGION PEAK CLADDING TEMPERATURE 1800 1600 -
140 50 1200-
- cD 1000- ,,.. .
E
ý1 Aj 400- \ 4 1 200O 0 50 100 150 200 250 300 350 Time After Break (s) 276688302 Figure 4 WCGS Limiting PCT Case - WCOBRAITRAC Peak Cladding Temperature for all 5 Rods
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 16 of 34 Wo f Creek ASTRUJM TCD BELOCA Ano ysis VESSEL SIDE BREAK FLOW PUMP SIDE BREAK FLOW cn E
0) cnf 0 50 100 150 200 250 300 Time After Break (s) 141182804 Figure 5 WCGS Limiting PCT Case - Break Flow
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 17 of 34 Wo f Creek A,STRUM TCD BELOCA An lysis PRESSUR IZER PRESSURE CL cn~
Vl)
P/
n 0 50 100 150 200 250 Time After Break (s) 141182804 Figure 6 WCGS Limiting PCT Case - Pressurizer Pressure
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 18 of 34 Wo f Creek A,STRUM TCD BELOCA Ana ysis LOOP 2 (INTACT LOOP) PUMP VOID FRACTION LOOP 1 BROKEN LOOP PUMP VOID FRACTION 0
~0 0 50 100 150 200 250 Time After Break (s) 141182804 Figure 7 WCGS Limiting PCT Case - Broken and Intact Loop Pump Void Fractions
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 19 of 34 Wo f Creek ASTRUM TCD BELOCA Ana ysis VAPOR FLOW RATE AT TOP OF CORE AVERGAE CHANNEL 13 I
cn E
0) 0 EL 0
0D_
0 10 20 30 40 50 Time After Breok (s) 141182804 Figure 8 WCGS Limiting PCT Case - Core Vapor Flow at the Top of the Core for a Core Average Channel
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 20 of 34 Wolf Creek ASTRUM TCD BELOCA Analysis VAPOR FLOW RATE AT TOP OF CORE HOT ASSEMBLY CHANNEL 15 16 14 12
~10 E
c-Q
_) 8 o 6 0
cj 4 0 10 20 30 40 50 Time After Break (s) 141182804 Figure 9 WCGS Limiting PCT Case - Core Vapor Flow at the Top of the Core for the Hot Assembly Channel
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 21 of 34 Wo f Creek ASTRUM TCD BELOCA Ana ysis LOWER PLENUM COLLAPSED LIQUID LEVEL 12 10 CL)
U)
M-J 0
0 50 100 150 200 250 300 350 Time After Break (s) 276688302 Figure 10 WCGS Limiting PCT Case - Lower Plenum Collapsed Liquid Level
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 22 of 34 Wolf Creek ASTRUM TCD BELOCA, nia ysis SINTACT LOOP 2 ACCUMULATOR MASS FLOW RATE E
a1) 0 Cn O0 0 50 100 150 200 250 Time After Break (s) 141182804 Figure 11 WCGS Limiting PCT Case - Intact Loop Accumulator Flow
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 23 of 34 Wo f Creek ASTRUM TCD BELOCA Ana ysis INTACT LOOP 2 SI MASS FLOW RATE (IHSI+RHR) 1 i1fl IfI 1 1000-800-600-400-Cfl 200-0-
I II I I I II I I I I I I I I I I LUU 0 50 100 150 200 250 Time After Breck (s) 141182804 Figure 12 WCGS Limiting PCT Case - Intact Loop (IHSI and RHR) Safety Injection Flow
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 24 of 34 Wof Creek ASTRUM TCD BELOCA Ana ysis INTACT LOOP 2 SI MASS FLOW RATE (HHS I) 100 80-60-40-Cn CD cjýý 20-0-
.20 0 50 100 150 200 250 Time After Break (s) 141182804 Figure 13 WCGS Limiting PCT Case - Intact Loop (EHHSI) Safety Injection Flow
Westinghouse Non-Proprietary Class 3 LTR-LIS- 13-557 Revision 0, NP-Attachment Page 25 of 34 Wo f Creek ASTRUM TCD BELOCA Ara ysis COLLAPSED LIQUID LEVEL N CORE AVERAGE CHANNEL 13 0) ci)
-J
-0) ci)
CJ)
-5 C-)
0 50 100 150 200 250 300 350 Time After Break (s) 276688302 Figure 14 WCGS Limiting PCT Case - Core Average Channel Collapsed Liquid Level I
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 26 of 34 Wolf Creek ASTRUM TCD BELOCA Anaoysis COLLAPSED LIQUID LEVEL IN INTACT LOOP 2 DOWNCOMER CT) 0 50 100 150 200 250 300 350 Time After Break (s) 276688302 Figure 15 WCGS Limiting PCT Case - Intact Loop Downcomer Collapsed Liquid Level
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 27 of 34 Wolf Creek ASTRUM TCD BELOCA An }ysis VESSEL LIQUID MASS 150000 E
cn 0 50 100 150 200 250 300 350 Time After Break (s) 276688302 Figure 16 WCGS Limiting PCT Case - Vessel Fluid Mass
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 28 of 34 Ca I cu I o ted Con t o i nmen t Bac k p r essu re (COCO)
Ana I ys i s Con to i nmen A Bockp ressu re (WC/T 40 0.) 3 0 . . . . . . .. . . . . . . ..
20 - .* . .
CD)
Cn2 cCO 0 50 100 150 200 250 300 2350 Time (s) 690332938 Figure 17 WCGS BE LBLOCA Analysis versus Calculated Containment Backpressure
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 29 of 34 PBOT/PMID Box For Wolf Creek BELOCA Project 0.50 0.31, 0.45 0.45 R
0 0.30,0.43 0.43,0.40 0.40 0
0 a.. 0.35 20 0.30 I I i I 0"325iI 0.27 I I I
. .. . -- -- - - - -L - - - - 0.26 0.25 --
- LI I 7 C
0.43, 0.23ý 0.20 - - -L-- - - - ---
0.15 i1 0.20 0.25 0.30 0.35 0.40 0.45 0.50 0.55 Power in Middle Third of Core (PMLD)
Figure 18 WCGS BE LBLOCA Analysis Axial Power Shape Operating Space Envelope
Westinghouse Non-Proprietary Class 3 LTR-LIS- 13-557 Revision 0, NP-Attachment Page 30 of 34 Section 8 (47.3")
OGL IL GT IGI OGL LPIGT OGL 45 3 Section 7 (22.2")
51 47 78 OmL IGL T IGL OGL LPIGT 06 41 4 40 440 2 4141 4 3 OGLP/SCOGL IGL L HA L L SC.. . .
Section 5 293.15 - (35.50)-
- E2:1]
278.15 -
20__ 20 1 ' "17 Section 4 goo% (13.22) 24- 0 19 264.93 - E!, ~ I29 '3 '2 '3 I~j 0 2 ' ~ 2 257.49 -
250.05 - - -r - - - T- -- ,- -
BB LIP GT HA UT SCIO "OP LP Be 239.74 -
229.45 - -
219.14 -
208.95 -
25 Section 3 198.74 - (144oý 0) 188.35 -
178.10 - - E1,1 167.85 -
iT -- - .-. ;
157.25
---- 2 146.65 -
13379 r" 120.93 -
Section 2 (73.06")
72.4 -.1 - L - 'TiEl 47.87 L ETI E] Channel 1(0 Gap 00 -
Figure 19 WCGS Vessel Model Noding Diagram (Vertical View)
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 31 of 34 SECTION 1: LOWER HEAD D] Channel ID#
Q Gap ID #
SECTION 2: LOWER PLENUM SECTION 3: CORE Figure 20 WCGS Vessel Sections 1 to 3 (Horizontal View)
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 32 of 34
.SECTION 4: CCFL REGION
[n Channel ID #
O GapID#
Loop 3 Loop 2 wlPressunzer 7 71 Loop 4 SECTION 5: UPPER PLENUM BELOW NOZZLES SECTION 6: NOZZLE REGION Figure 21 WCGS Vessel Sections 4 to 6 (Horizontal View)
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 33 of 34 SECTION 7: UPPER PLENUM ABOVE NOZZLES SECTION 8: UPPER HEAD n Channel ID #
O Gap ID #
SECTION 9: UPPER HEAD SECTION 10: UPPER HEAD Figure 22 WCGS Vessel Sections 7 to 10 (Horizontal View)
Westinghouse Non-Proprietary Class 3 LTR-LIS-13-557 Revision 0, NP-Attachment Page 34 of 34 VALVE3 VALVE3 VALVE3 Figure 23 WCGS WCOBRA/TRAC Model Loop Layout
Enclosure III to ET 14-0008 Westinghouse Electric Company LLC, CAW-14-3879, "Application for Withholding Proprietary Information from Public Disclosure" (7 pages)
Westinghouse Electric Company
ý8P Westinghouse Engineering, Equipment and Major Projects 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 720-0754 11555 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 Proj letter: SAP-14-16 CAW- 14-3879 January 21, 2014 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
Subject:
LTR-LIS-13-557 P-Attachment, Revision 0, "Application of Westinghouse Best-Estimate Large Break LOCA Methodology to the Wolf Creek Generating Station" (Proprietary)
The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-14-3879 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.
Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Wolf Creek Nuclear Operating Corporation (WCNOC).
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse Affidavit should reference CAW-] 4-3879, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 310, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.
Very truly yours, James A. Gresham, Manager Regulatory Compliance Enclosures
CAW-14-3879 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
ss COUNTY OF BUTLER:
Before me, the undersigned authority, personally appeared James A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
,lames A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this __ _ day of, natca 2014 NotarPublic COMMONWEALTH OF PENNSYtVANIA
{ ~Notarlial Sig I[ Renee GIampole, Notary Public Penn Twp., Westmoreland County
. My Commission Eixpires Sept. 25, 2017 MEMBER, PENNSYLVANIAASOaATIO OF NOp ,rAJE
2 CAW-14-3879 (1) 1 am Manager, Regulatory Compliance, in Engineering, Equipment and Major Projects, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.
(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.
(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.
(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of
3 CAW-14-3879 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.
(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f) It contains patentable ideas, for which patent protection may be desirable.
(iii) There are sound policy reasons behind the Westinghouse system which include the following:
(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.
(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
4 CAW-14-3879 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.
(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.
(iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.
(v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.
(vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in LTR-LIS-1 3-557 P-Attachment, Revision 0, "Application of Westinghouse Best-Estimate Large Break LOCA Methodology to the Wolf Creek Generating Station" (Proprietary), for submittal to the Commission, being transmitted by Wolf Creek Nuclear Operating Corporation (WCNOC) letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with the results of and method for the thermal conductivity degradation evaluation for Wolf Creek, and may be used only for that purpose.
5 CAW-1 4-3879 (a) This information is part of that which will enable Westinghouse to:
(i) Provide input to WCNOC to provide to the U.S. Nuclear Regulatory Commission regarding the evaluation of thermal conductivity degradation.
(ii) Provide licensing support for customer submittal.
(b) Further this information has substantial commercial value as follows:
(i) Westinghouse plans to sell the use of the information to its customers for the purpose of addressing the effects of thermal conductivity degradation.
(ii) Westinghouse can sell support and defense of the technology to its customer in the licensing process.
(iii) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar calculations and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.
Further the deponent sayeth not.
PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.
In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).
COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.